millstone, units 1, 2 & 3 - 10 cfr 50.59 and commitment change … · 2012. 11. 21. · serial...

DominionDominion Nuclear Connecticut, Inc.Millstone Power StationRope Ferry RoadWaterford, CT 06385

JUN 2 8 2006U.S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555

Serial No.MPS Lic/GJCDocket Nos.

License Nos.

06-514RO50-24550-33650-423DPR-21DPR-65NPF-49

DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNITS 1, 2 AND 310 CFR 50.59 AND COMMITMENT CHANGE REPORT FOR 2005

Pursuant to the provisions of 10 CFR 50.59(d)(2), the reports for changes made to thefacility for Millstone Power Station Units 2 and 3 (MPS 2 and 3) are submitted viaAttachments 1 and 2 respectively. Attachment 3 reports changes made common to allMillstone Power Station units. There were no changes made to the facility for MillstonePower Station Unit 1 (MPS 1).

Additionally, during 2005, there were no commitment changes for MPS 1, 2, or 3. Thisconstitutes the annual Commitment Change Report consistent with the Millstone PowerStation's Regulatory Commitment Management Program.

If you have any questions or require additional information, please contact Mr. David W.Dodson at (860) 447-1791, extension 2346.

Very truly yours,

J. la PriceSit rice Pres~ident - Millstone

•c-7j--5'(

Serial No. 06-514.10 CFR 50.59 and Commitment Change Report for 2005

of 2Attachments: 3

Commitments made in this letter: None.

cc: U.S. Nuclear Regulatory CommissionRegion I475 Allendale RoadKing of Prussia, PA 19406-1415

Mr. Joseph A. GollaNRC Project Manager- Millstone Unit 1U.S. Nuclear Regulatory Commission, Mail Stop 012H2One White Flint North11555 Rockville PikeRockville, MD 20852-2738

Mr. R. PrinceNRC InspectorU.S. Nuclear Regulatory CommissionRegion I475 Allendale RoadKing of Prussia, PA 19406-1415

Mr. V. NersesSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North11555 Rockville PikeMail Stop 8C2Rockville, MD 20852-2738

Mr. S. M. SchneiderNRC Senior Resident InspectorMillstone Power Station

Attachment I

10 CFR 50.59 REPORT FOR 2005

Millstone Power Station Unit 2Dominion Nuclear Connecticut, Inc. (DNC)

Serial No. 06-51410 CFR 50.59 and Commitment Change Report for 2005

Attachment 1/ Page 1

$2-EV-01-0023Added Disconnect Switches to Appendix R MOVs 2-MS-65A, 2-MS-65B, and 2-MS-202Affected Documents:LBDCR 05-MP2-010

Description

Design Change Request (DCR) M2-00030 modified the Appendix R MOVs 2-MS-65A, 2-MS-65B, and 2-MS-202 to eliminate an Appendix R requirement to remove starter coils from theirapplicable Motor Control Center (MCC) circuit breakers, and installed new keyed disconnectswitches instead. Condition Report (CR) CR-05-05500 identified that the Millstone Unit 2(MP2) Technical Requirements Manual (TRM) operability requirement in Table 7.1.17-1 forvalve 2-MS-65B did not reflect this modification.

Reason

Investigation into the CR identified that the physical work for 2-MS-65B was performedseparately from valves 2-MS-65A and 2-MS-202. The associated TRM update was thereforesplit into two partial submittals. Upon completion of the DCR, the change to the TRM wasissued, but a subsequent change inadvertently deleted the revised requirement. ThisLicense Basis Document Change Request (LBDCR) restored the correct wording.

Summary

Neither the original change, nor this administrative correction, adversely affect equipmentdesign or operation, and no changes were made to Technical Specification required safetylimits or safety system settings. The changes did not adversely affect any accident mitigationfunction and are unrelated to any evaluation methodology as described in the Final SafetyAnalysis Report (FSAR). The TRM change did not result in a design basis limit for a fissionproduct barrier as described in the FSAR being exceeded or altered; does not increase thefrequency of occurrence or the consequences of an accident previously evaluated in theFSAR; and does not increase the likelihood of occurrence of a malfunction to a system,structure or component important to safety.



$2-EV-05-0002Reload Design for Millstone Unit 2 Cycle 17FSARCR for Parent Document DCR No. M2-04004, Rev. 0Affected Documents:DCR No. M2-04004, Revision 0FSARCR No. 05-MP2-009, Revision 0

Description

This design change record (DCR) evaluates the Millstone Unit 2 (MP2) Cycle 17 core designcharacteristics in order to demonstrate that the Cycle 17 core design and operation areacceptable. The Cycle 17 core was fueled with 68 fresh MIB-8 fuel assemblies, 68 once-burned MIB-7 fuel assemblies, 80 twice-bumed MIB-6 fuel assemblies, and 1 twice burnedANF-2 fuel assembly.

Two (2) minor fuel mechanical design changes were implemented for the first time inCycle 17 with the installation of MIB-8 fuel: Upset Shape Welding of fuel rod upper andlower end caps changed the method used to weld the end caps to the cladding tube,and the hole in the center location of the Lower Tie Plate (LTP) was drilledapproximately 0.6 inch deeper to minimize impact due to irradiation-related growth ofthe incore instrumentation (ICI) guide tubes.

The Final Safety Analysis Report Change Request (FSARCR) documents relatedchanges to the FSAR that provide clarification, document editorial changes, and reflectCycle 17 specific analyses.

Reason

This change was necessary to allow Cycle 17 operation and the continued production ofelectricity by Millstone Unit 2.

Summary

This evaluation was necessary due to the Cycle 17 specific analysis for pre-scram mainsteam line break (MSLB) event with power available (FSAR Section 14.1.5.1.6.1). In allcases, the Cycle 17 specific events remain bounded by the analyses of record,however, the Cycle 17 specific result (i.e., fuel rod failures due to exceeding theDeparture from Nucleate Boiling Ratio limit) is more limiting than that for Cycle 16,although still bounded by the Cycle 15 analysis of record. The Cycle 17 specific resultfor this event does not result in more than a minimal increase in consequences of theaccidents evaluated in the FSAR. The other changes made by this DCR have noimpact on these consequences.

The core design is essentially unchanged from the previous cycle with regard toanalysis assumptions and the safety analyses of record. The components affected bythe two (2) changes in fuel design do not relate to the initiation of FSAR Chapter 14events. As such, the changes described in the DCR will not affect the frequency ofoccurrence of the FSAR events. This evaluation found no issues or events that



indicated the possibility of a new, unanalyzed event, nor created the possibility of anaccident of a different type than those previously evaluated in the FSAR.

Attachment 2


Millstone Power Station Unit 3Dominion Nuclear Connecticut, Inc. (DNC)



S3-EV-98-0201Implementation of Permanent Chemical Transfer System

Affected Documents:FSARCR 98-MP3-126

Description

Design Change Request M3-97018 made permanent the use of ethanolamine (ETA) as thepH agent at Millstone Unit 3 (MP3), and retired a portable storage tank and connections witha permanently installed chemical addition system. Final Safety Analysis Report (FSAR)Figure 10.3-03, Sheet 1A, illustrated the temporary modification.

Reason

FSAR Change Request 98-MP3-126 modified FSAR section 10 to reflect the deletion ofFigure 10.3-03, Sheet 1A, and associated references to this drawing.

Summary

Deleting this Figure and its attendant references provides for conformity between designbasis documents and the actual configuration of the chemical addition system. No newmodifications were made as a result of correcting this historical inaccuracy in the FSAR.



S3-EV-02-0020Correction to Text in a TRM Surveillance Requirement and its Bases for the TRMContainment Structural Integrity LCOAffected Documents:LBDCR 05-MP3-006

Description

Condition Report (CR) CR-04-07025 identified that Millstone Unit 3 (MP3) TechnicalRequirements Manual (TRM) Surveillance Requirement (SR) 4.6.1.6.3.c was mis-worded,and that the Bases inappropriately implied that entry into a Technical Specification ActionStatement might be deferred by eight (8) hours.

The original LBDCR revised the MP3 TRM to include an optional, variable Allowed OutageTime (AOT) for the Engineered Safety Features (ESF) Building Porous ConcreteGroundwater Sump Pump (3SRW-P5), based upon groundwater inleakage rate. A newsurveillance requirement was added to verify groundwater inleakage rate every 8 hours whenoperating within the constraints of the variable AOT. SR 4.6.1.6.3.c amended the inleakageflowrate requirement from less than 2209 gallons per day (gpd) to less than or equal to 2209gpd for mathematical accuracy. The changes made by License Basis Document ChangeRequest (LBDCR) 05-MP3-006 corrected the wording of the SR and aligned it with theoriginal intended requirements for use of the variable AOT, as established in LBDCR 02-3-13.

Reason

This License Basis Document Change Request (LBDCR) provided correction to the text inthe TRM Surveillance Requirement 4.6.1.6.3.c and its Bases. The intended requirementsassociated with the use of the variable AOT calculation in Action a.2, and the correspondinguse of TRM SR 4.6.1.6.3.c to implement this calculation were both unaffected and entirelyconsistent with the previous approval for their use in the TRM.

Summary

Neither the original change, nor this administrative correction, adversely affect equipmentdesign or operation, and no changes were made to Technical Specification required safetylimits or safety system settings. The changes did not adversely affect any accident mitigationfunction and are unrelated to any evaluation methodology as described in the Final SafetyAnalysis Report (FSAR). The TRM change did not result in a design basis limit for a fissionproduct barrier as described in the FSAR being exceeded or altered; does not increase thefrequency of occurrence or the consequences of an accident previously evaluated in theFSAR; and does not increase the likelihood of occurrence of a malfunction to a system,structure or component important to safety.



$3-EV-05-0001Reload Design for Millstone Unit 3 Cycle 11Affected Documents:DCR No. M3-05001, Revision ALBDCR No. 05-MP3-008FSARCR No. 05-MP3-019

Description

Design Change Request (DCR) M3-05001, Revision A documents and evaluates theMillstone Unit 3 (MP3) Cycle 11 core for MODES 5 and 6 conditions. As per the MillstoneDesign Control Manual (DCM), Revision A is designated as an "Early Approval forImplementation" DCR. The approval of this DCR allows outage work such as core reload toprogress, up to and including MODE 5 conditions, while other documentation related to thefinal (Revision 0) DCR is in preparation. The final (Revision 0) DCR subsequently supportedall MODES of Cycle 11 operation.

The Cycle 11 core is fueled with 73 fresh Batch N RFA-2 fuel assemblies, 64 once burnedBatch M RFA-2 fuel assemblies, 8 once burned Batch M NGF LTA fuel assemblies, and 48twice burned Batch L RFA fuel assemblies. For Cycle 11, an increase in fuel cycle installedcapacity resulted in a slight increase in boration requirements to the point where the statedminimum indicated Boric Acid Storage System volume was insufficient to satisfy MP3Technical Requirements Manual (TRM) 3.1.2.6 requirements. The Cycle 11 boron designrequirement analysis from the fuel vendor identified the need to increase the minimum BATsystem volume from a value of 21,802 gallons (usable) to at least 22,095 gallons (usable).The revision made to TRM LCO 3.1.2.6.a (LBDCR No. 05-MP3-008), increased the minimumindicated borated water volume to 32,000 gallons, to bound future core designs. Changesmade to the MP3 Final Safety Analysis Report (FSAR) included a related change to Table9.3-4, Cycle 11 specific information in Chapters 4 and 15, and minor editorial changes inChapters 4 and 15.

Reason

This change was necessary to allow Cycle 11 operation and the continued production ofelectricity by Millstone Unit 3.

Summary

This evaluation only addressed Revision A to DCR No. M3-05001, which was applicable toCycle 11, MODES 5 and 6. However, the evaluation also addressed the related LBDCR andFSARCR for all MODES of Cycle 11 operation. The TRM that was revised was onlyapplicable in MODES 1-4, however, the COLR revisions needed to be implemented beforethen to support Cycle 11 MODE 6.

The implementation of the MP3 Cycle 11 reload core design does not affect any accidents ormalfunctions evaluated in the FSAR, nor does it create a new type of event not previouslyevaluated in the FSAR. There is no negative impact on any fission product barrier asdescribed in the FSAR due to the implementation of the Cycle 11 reload core design. Thereload core design criteria and licensing basis acceptance criteria evaluations do not result in



a departure from any evaluation methodology used in establishing the Millstone Unit 3 designbasis or safety analysis.



S3-EV-05-0002Reload Design for Millstone Unit 3 Cycle 11Affected Documents:DCR No. M3-05001, Revision 0

Description

Design Change Request (DCR) No. M3-05001, Revision 0 was written to evaluate all modes(MODES 1 to 6) of Millstone Unit 3 (MP3) Cycle 11 core design and operation. Thepreviously issued DCR, M3-05001, Revision A (Early Approval for Implementation), wasevaluated in S3-EV-05-0001 and allowed Cycle 11 operation in MODES 5 and 6 only. ThisRevision 0 DCR (M3-05001) documents the loading of 73 new fuel assemblies for Cycle 11 ina low leakage loading pattern with 72 once-burned assemblies from Cycle 10 and 48 twice-burned assemblies used in the last two cycles. The feed assemblies are of the RFA-2design.

Reason

This change was necessary in order to allow Cycle 11 operation and the continuedproduction of electricity by Millstone Unit 3.

Summary

The Cycle 11 core design does not adversely affect any accidents or malfunctions evaluatedin the Final Safety Analysis Report (FSAR), nor create a new type of event not previouslyevaluated in the FSAR. There is no negative impact on any fission product barrier. Themethods used for designing the Cycle 11 core do not deviate from those used in establishingthe MP3 design basis or safety analyses.



S3-EV-05-0003Turbine Overspeed Protection Maintenance and Testing Program ManualAffected Documents:U3-24-TOP-PRG, Revision 001-01

Description

A one-time change was made to the Turbine Overspeed Protection (TOP) program, allowinga relaxation of the requirement to perform a surface examination of at least one stop valve onan approximately 3.33 year frequency to an approximately 4.5 year frequency.

Reason

U3-24-TOP-PRG required examinations ensured that at approximately 3.33 year intervals,at least one main steam stop valve, one main steam control valve, and one main reheatvalve and one reheat intercept valve would be dismantled, and the valve seats, discs, andstems inspected visually, and by surface examination using magnetic particleexaminations. During the last refueling outage (3R10), one stop valve, one control valveand a combined intercept valve were disassembled and inspected, however the surfaceexaminations were not performed. (This deviation was documented in CR-05-11552.) Theexaminations performed in 3R09 did not include a main steam stop valve, therefore,without further examinations, the program inspection requirement for surface examinationsof the main steam stop valves every 3.33 years will not be met. This deviation from therequirements of the program is deemed to have an adverse effect on meeting the designrequirements of General Design Criteria 4. Therefore this evaluation addresses the changeto the program requirements and evaluates allowing the main stop valve surfaceexamination to be extended to the next refueling outage.

Summary

The missed magnetic particle examinations of the three main turbine valve stems do notincrease the chance of a valve failure and do not increase the probability of a turbineoverspeed event. With the missed examinations, only one of the three valve typeschallenges the Final Safety Analysis Report (FSAR) stated frequency for requiredinspections. The current disassemblies and inspections of the valves, coupled with thoseperformed in the past are consistent with or exceed the FSAR requirements for the mainturbine valves. Past performance of the disassemblies and inspections of these particularvalves, as well as the frequency of performance of the disassemblies of the remainder ofthe turbine valves continues to be consistent with the Turbine Overspeed ProtectionProgram Manual. Valve operation and functionality are not challenged by the missedexam.

The change in frequency of surface examinations of the Main Steam Stop Valves from 3.33years to approximately 4.5 years does not affect the probability of, or consequences ofaccidents described in the FSAR. It does not result in more than a minimal increase in theprobability of malfunction of the turbine overspeed protection, nor does it affect the



consequences of malfunction of any SSC important to safety. The change does not affectany method of evaluation described in the FSAR, or create a new event or a malfunctionwith a different result. The change has no effect on any fission product boundaries.

Attachment 3


Millstone Power Station Units 1, 2, & 3Dominion Nuclear Connecticut, Inc. (DNC)



SG-EV-05-0001Design, Installation and Operation of an Independent Spent Fuel Storage Installation(ISFSI) for Millstone StationAffected Documents:DCR MG-04010

Description

This evaluation was written to support the design, installation, and operation of an ISFSI forMillstone Station. Included in these activities was the selection of Millstone Unit 2 (MP2)spent fuel to be loaded in Dry Shielded Canisters (DSC) in the cask laydown area of theMP2 spent fuel pool (SFP), DSC transport in a shielded transfer cask between the MP2SFP and the ISFSI site, and DSC placement into and removal from a Horizontal StorageModule (HSM). Millstone had selected the NUHOMS® spent fuel storage system underTransnuclear Corporation's general license, as authorized per 10 CFR 72, and approvedby the U.S. Nuclear Regulatory Commission in Certificate of Compliance Number 1004.

Reason

The ISFSI was designed to provide a "dry" option for spent fuel storage at the MillstonePower Station.

Summary

The subject Design Change Record (DCR) was prepared based on the NUHOMS® FinalSafety Analysis Report (FSAR) and Technical Specifications, and comprises the initialMillstone ISFSI licensing and design bases. Therefore, evaluation under 10 CFR 50.59criterion is appropriate. Changes made to the ISFSI after its initial release toOperations will be evaluated using the 10 CFR 72.48 Screen/Evaluation process.

Since the initial fuel loading campaign for the ISFSI involved only MP2, this evaluationwas limited to the 50.59 licensing and design basis requirements for dry fuel storage asthey apply to MP2.

Handling of the spent fuel during DSC loading activities remains essentially unchangedfrom the normal MP2 fuel handling process. The spent fuel cooling function is notadversely affected by the fuel load activities. On February 15, 2005, Dominion NuclearConnecticut, Inc. received approval for a limited exemption from the requirements of10 CFR 50.68 for criticality control measures with fuel loaded in DSCs in the SFP.Commitments made in support of the exemption request ensured consistent boronconcentration in the cask laydown area.

Operation of the Millstone ISFSI for storing MP2 fuel cannot involve a change thatadversely affects a MP2 FSAR described design function; adversely affects how thesedesign functions are performed or controlled; utilizes a revised or alternate evaluationmethodology than that used in establishing design bases or used in the safety analysis;constitutes a test or experiment not described in the MP2 FSAR where a system,structure, or component is utilized or controlled in a manner that is outside the boundsof the design or is inconsistent with analyses or descriptions used in the MP2 FSAR.

Events postulated by the NUHOMS® FSAR were outside the scope of this 50.59 evaluation.